Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems
Section snippets
Background
Ferritic steels have been considered as candidate structural materials for fusion power plants since the late 1970s. The reason being is that the data obtained from fast reactor irradiation shows that ferritic steels are more swelling resistant than austenitic stainless steels. Moreover, their higher thermal conductivity and lower thermal expansion coefficients lead to improved resistance to thermal stresses characteristic for a fusion power plant operating in a pulsed mode.
Mainly high Cr
Utilised data and assumptions for their comparison
Critical sets of surveillance PWR data are used for this study. Representative clean ferritic western type PWR RPV steels [22] are selected. The upper fluence is approximately corresponding to 80–90 mdpa (milli-dpa). Additionally, surveillance data at ∼270 °C taken from Russian type pressurised water reactor WWER-440 type are also considered in the study. An important remark should be made that only ‘clean’ steels with sufficiently low Cu and P content are utilised. The fluence range for
Additional dataset for comparison
Very comprehensive research has been done on using different Cr content steels for fusion applications. Mainly modified 2.25%Cr and higher Cr steels (up to 12%Cr) have been considered as potential candidates [1], [2], [3], [4], [5], [6], [7], [8], [9]. High dose embrittlement data obtained at 365 °C irradiation in Fast Flux Test Facility (FFTF), owned by the U.S. Department of Energy, are available in the literature [9]. The data and their trend are used for comparison in relation to the
Data comparison and modelling implications
Since relatively clean steels were considered for the study, only basic matrix damage (material hardening occurring during irradiation) was taken into account in the embrittlement process. Embrittlement data on low Cr steels PWR, BWR and WWER 440 steels, and high Cr (up to 9%Cr) Eurofer 97 and 9Cr2WVTa steels, represented as an increase in the tensile yield stress as a function of the dose are shown in Fig. 1. For better representation a logarithmic scale was used. Based on a physically sound
Chromium effect on irradiation stability
Embrittlement data, expressed again in tensile yield stress increase as a function of the dose, at higher doses (up to 28 dpa) and higher irradiation temperature of 365 °C for different Cr content steels are analysed and compared in Fig. 2 to the established embrittlement trend shown in Fig. 1. Evidence for chromium stabilisation of irradiation damage in RPV weld metals is investigated and shown in Ref. [36]. The measured DBTT shift for high-Cr welds lie well below the value measured for low Cr
Conclusions
Manufacturing of extremely large and heavy pressure vessels using expensive non-conventional material and complicated joining procedures is not sustainable whereas the utilisation of the best conventional well proven materials, like LWR RPV steels may offer significant advantages. Using conventional low Cr clean RPV steel might be a smart, economical and feasible solution for pressure vessel and other massive components of the future nuclear reactors.
Surveillance data from WWER-440 materials up
References (36)
- et al.
Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation
J Nucl Mater
(1988) - et al.
Impact behavior of 9Cr-1MoVNb and 12Cr-1MoVW steels
J Nucl Mater
(1991) Chromium–molybdenum steels for fusion reactor first Walls – a review
Nucl Eng Des
(1982)- et al.
Development of low-Cr, Cr-W steels for fusion
J Nucl Mater
(1995) - et al.
Effect of heat treatment and irradiation temperature on impact properties of Cr-W-V ferritic steels
J Nucl Mater
(1999) - et al.
Impact behavior of reduced-activation steels irradiated to 24 dpa
J Nucl Mater
(1996) Creep-rupture behaviour of a bainitic 2.25Cr1Mo steel
Int J Press Vessels Pip
(1980)- et al.
Role of nickel in a semi-mechanistic analytical model for radiation embrittlement of model alloys
J Nucl Mater
(2005) - et al.
Primary damage formation in BCC iron
J Nucl Mater
(1997) - et al.
Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor
J Nucl Mater
(2011)
Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors
J Nucl Mater
Effect of irradiation temperature in PWR RPV materials and its inclusion in semi-mechanistic model
Scr Mater
Specification of stress limits for irradiated 316L(N)-IG steel in ITER structural design criteria
J Nucl Mater
Structural materials for Gen-IV nuclear reactors: challenges and opportunities
J Nucl Mater
Evidence for chromium stabilisation of irradiation damage in RPV weld metals
Int J Press Vessels Pip
Tensile behavior of three commercial ferritic steels after low-temperature irradiation, Proceedings of the topical conference on ferritic alloys for use in nuclear energy technologies
A comparison of low-chromium and high-chromium reduced-activation steels for fusion applications
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Computational Kinetics: Application to Nuclear Materials
2020, Comprehensive Nuclear Materials: Second EditionMicrostructural characterization of the synergic effects of dynamic strain ageing and hydrogen on fracture behaviour of low-alloy RPV steels in high-temperature water environments
2020, Materials CharacterizationCitation Excerpt :Commercial RPVs are made of low-alloy steels (LAS) with typical upper bainite microstructure, such as MnMoNi and NiMoCr steel grades [1]. Irradiation embrittlement [2–5], as well as the high-temperature water (HTW) environment and the absorption of hydrogen from the reactor coolant and corrosion reactions can lead to the reduction of the fracture resistance of RPV. The degradation effects of HTW environment and hydrogen on the fracture properties of RPV steels can be concomitant with the effects of irradiation embrittlement [6], dynamic strain ageing (DSA) [7,8] or environmentally-assisted cracking (EAC) [9–11].
On the evolution of mechanical properties and microstructure of ferritic-bainitic (FB) 2.25Cr-1Mo (Grade 22) steel during high-temperature creep
2020, MaterialiaCitation Excerpt :In a typical application, Grade 22 steel is used for superheater tubing as well as a filler metal for joining steam piping. Due to its excellent creep-rupture strength [4,5] and radiation resistance [6,7], it is also considered as a candidate material for application in the next generation (Generation IV) of nuclear reactors [8] as a pressure vessel material [9]. Because of its use at high temperatures under loading, it is of technological importance to monitor creep damage of components made of Grade 22 steel in order to determine the remaining creep life of a component under operation conditions.
Environmental degradation of fracture resistance in high-temperature water environments of low-alloy reactor pressure vessel steels with high sulphur or phosphorus contents
2019, Corrosion ScienceCitation Excerpt :The body of the RPV is made of fine-grained low alloy steel (LAS) with granular upper bainite microstructure, such as MnMoNi or NiMoCr steel series [2,3]. The fracture resistance of RPV is mainly degraded by irradiation embrittlement [4–6] due to neutrons and thermal ageing [7], and potentially by the high-temperature water (HTW) environment and the absorption of hydrogen. The hydrogen absorption from the reactor coolant is due to radiolysis, intentional additions or corrosion reactions.
Effect of dynamic strain ageing on environmental degradation of fracture resistance of low-alloy RPV steels in high-temperature water environments
2019, Corrosion ScienceCitation Excerpt :Commercial RPVs are made of fine-grained low-alloy steels with granular upper bainite microstructure, such as MnMoNi and NiMoCr steel grades [1]. The fracture resistance of RPV can be reduced by irradiation embrittlement due to neutrons and thermal ageing [2–4], as well as potentially by the high-temperature water (HTW) environment and the absorption of hydrogen from the reactor coolant and corrosion reactions. The fracture behaviour of RPV steels in HTW is expected to be affected by material (strength, chemical composition, DSA susceptibility, microstructure, etc.), environmental (temperature, water chemistry, etc.) and mechanical loading parameters (strain rate, constraints, etc.).
Motivation for utilizing new high-performance advanced materials in nuclear energy systems
2016, Current Opinion in Solid State and Materials ScienceCitation Excerpt :This reduction in thickness, and therefore ingot mass, would greatly reduce fabrication costs while enabling access to numerous smaller capacity vendors (existing in the US and elsewhere). Furthermore, safety and reliability may be enhanced for this critical component given the anticipated improved irradiation resistance in these alloys [42–44]. For example, the unirradiated fracture toughness for 3Cr-3WV steel is similar or superior to conventional 1Cr RPV steels [40], and the irradiation-induced shift in ductile to brittle transition temperature and reduction in upper shelf toughness is expected to be markedly less for the 3Cr-3WV steels due to reduced radiation hardening [10,11], although additional experimental data are needed to confirm this expectation.